1. Field of the Invention
The present invention relates to zirconium-base alloys for nuclear reactors and to nuclear reactor components comprising the zirconium-base alloys.
2. Background Art
Zirconium alloys are now extensively used as fuel cladding materials and core structural materials in boiling-water reactors, pressurized-water reactors, and so forth. The most commonly employed so far are Zircaloy-2 containing 1.2-1.7 wt. % Sn, 0.07-0.20 wt. % Fe, 0.05-0.15 wt. % Cr, 0.03-0.08 wt. % Ni, and the balance being Zr, and Zircaloy-4 containing 1.2-1.7 wt. % Sn, 0.18-0.24 wt. % Fe, 0.07-0.13 wt. % Cr, and the balance being Zr. In addition to these zirconium alloys, Zrxe2x88x922.5% Nb alloy, Zrxe2x88x921% Nb alloy, etc. are used for nuclear reactors as well. The above-described zirconium alloys have been developed in view of neutron economy, strength and corrosion resistance.
One problem with boiling-water reactors is that corrosion called nodular corrosion occurs, during operation, at localized regions on the surfaces of the above-described materials to give lenticular corrosion layers. The nodular corrosion process proceeds with the progress of irradiation, and the corrosion layers can come off when they got thick. Thus, if nodular corrosion occurs, structural materials undergo decrease in thickness. Moreover, since the falling of corrosion layers increases the radioactivity in a coolant, there is such a fear that the exposure dose during periodic inspection becomes higher. Known techniques for preventing occurrence of nodular corrosion include a heat treatment method in which an alloy is heated, for a short period of time, at a temperature in the temperature range in which the alloy exists in xcex1+xcex2 or xcex2 phase and is then rapidly quenched (Japanese Patent Publications Nos. 45699/1986 and No. 58223/1988), and a method in which the composition of an alloy is modified (Japanese Laid-Open Patent Publications No. 43450/1985 and No. 228442/1987).
For example, in the process of heat treatment in which a zircaloy once heated to a high temperature in the temperature range in which the alloy exists in xcex1+xcex2 or xcex2 phase is rapidly quenched (xcex1+xcex2-quenching or xcex2-quenching), intermetallic compounds present in the alloy as precipitates (mainly, two types of compounds Zr(Cr, Fe)2 and Zr2(Ni, Fe) are present) are fine. It is known that since the precipitates are fine in the course of xcex1+xcex2 or xcex2 quenching, occurrence of nodular corrosion is prevented to greatly improve the corrosion resistance of the alloy.
Occurrence of nodular corrosion is successfully prevented by the above-described means, and the type of corrosion is now being shifted to uniform corrosion in which an oxide film, corrosion product, grows uniformly.
The above-described materials having improved resistance to nodular corrosion tend to uniformly corrode to give, on their surfaces, uniform oxide films, which are extremely thin and show excellent protective properties. It can, therefore, be said that these materials satisfactorily fulfill their functions under the conditions of operation of the current nuclear reactors.
For the purpose of improving the economical efficiency of the current nuclear power plants, on the other hand, efforts are now being made step by step to further increase fuel burn-up. If the in-core fuel dwelling time becomes longer with the increase in fuel burn-up, alloys to be used for reactor components are required to have not only corrosion resistance but also proper hydrogen absorption characteristics. Namely, such a possibility is pointed out that if the in-core fuel dwelling time becomes longer, zirconium-alloy-based materials absorb increased amounts of hydrogen and structural materials made from such materials are embrittled.
As described in Japanese Laid-Open Patent Publication No. 228442/1987, it is qualitatively known that zircaloys having higher Fe contents absorb decreased amounts of hydrogen. In the light of this fact, Zircaloy-2, which is now mainly used for boiling-water reactors, has been modified, for example, by increasing its Fe content within the specified range. The modified Zircaloy-2 well functions under the current operating conditions when viewed from corrosion resistance and hydrogen absorption characteristics. However, due to such a change in circumstances that hydrogen is fed to the coolant water supply system of a nuclear reactor and due to a demand for decrease in the wall thickness of fuel structural materials, the hydrogen absorption requirements, in particular, are becoming severer in recent years. To lower hydrogen absorption is, therefore, an important matter for improvement in the reliability and performance of nuclear reactor fuels.
To meet the long-term use of nuclear reactors involving the attainment of high burn-up and to cope with such a change in circumstances as feeding of hydrogen and so on, it is urgently demanded to develop zirconium alloys that have high corrosion resistance and less absorb hydrogen.
The present invention relates to zirconium alloys for use in nuclear reactors. An object of the present invention is to provide zirconium-base alloys that show high corrosion resistance and low hydrogen absorption even when they are used, under the core conditions, in light-water-cooled nuclear power reactors for a prolonged period of time, and nuclear reactor components comprising the zirconium-base alloys.
A zirconium-base alloy for nuclear reactors according to the present invention comprises 0.5-2 wt. % Sn, 0.07-0.6 wt. % Fe, 0.03-0.2 wt. % Ni, 0.05-0.2 wt. % Cr, and the balance being zirconium and unavoidable impurities, wherein the Fe content (X wt. %) of the zirconium-base alloy and the mean size (Y nm) of precipitates in the zirconium-base alloy are present in a region on the x (Fe content X) and y (mean precipitate size) rectangular coordinates, surrounded by the following five lines:
i) Y=xe2x88x92444xc3x97X+154,
ii) Y=910xc3x97Xxe2x88x9246,
iii) Y=0,
iv) Y=300, and
v) X=0.6.
A preferable zirconium-base alloy for nuclear reactors according to the present invention is one in which the above-described equation i) is Y=xe2x88x92989xc3x97X+362.
Further, a zirconium-base alloy for nuclear reactors according to the present invention comprises 0.5-2 wt. % Sn, 0.07-0.6 wt. % Fe, 0.03-0.2 wt. % Ni, 0.05-0.2 wt. % Cr, and the balance being zirconium and unavoidable impurities, wherein the Fe content (X wt. %) and the annealing parameter (xcexa3Ai) of the zirconium-base alloy are present in a region on the x (Fe content X) and y (annealing parameter) rectangular coordinates, surrounded by the following five lines:
i) 30+1.6xc3x97107xc3x97exp (0.7xc3x97log (Y))=xe2x88x92444xc3x97X+154,
ii) 30+1.6xc3x97107xc3x97exp (0.7xc3x97log (Y))=910xc3x97Xxe2x88x9246,
iii) Y=1xc3x9710xe2x88x9221,
iv) Y=1xc3x9710xe2x88x9215, and
v) X=0.6.
A preferable zirconium-base alloy for nuclear reactors according to the present invention is one in which the above-described equation i) is 30+1.6xc3x97107xc3x97exp (0.7xc3x97log (Y))=xe2x88x92989xc3x97X+362.
The mean size (Y nm) of precipitates in a zirconium-base alloy can be determined, for example, by the following method: a zirconium-base alloy is made into a thin film, and precipitates contained in this thin film are observed by a transmission electron microscope; the images of the precipitates are recorded as image data; and these data are analyzed to determine the diameters of circles approximate to the precipitates and the mean diameter is then obtained by calculation. Another method useful for determining the mean precipitate size is as follows: a zirconium-base alloy is chemically etched, and precipitates on the etched alloy surface are observed by a scanning electron microscope. In general, approximately 200 precipitates are needed to obtain stable size distribution, so that it is desirable to observe 200 or more precipitates.
The annealing parameter (xcexa3Ai) is an index of the amount of heat a zirconium-base alloy gains during heat treatment, such as annealing, that is conducted after subjecting the alloy to xcex2 or xcex1+xcex2-quenching. The annealing parameter can be quantified by using the following equation:
xcexa3Ai=xcexa3tixc3x97exp(xe2x88x9240,000/Ti)
wherein ti is a retention time (h) at a heat treatment temperature Ti, and Ti is a heat treatment temperature (K). The xcex2-quenching herein means that a zirconium-base alloy is heated to a temperature in the temperature range in which the alloy exists in xcex2 phase and is then rapidly quenched, and the xcex1+xcex2-quenching means that a zirconium-base alloy is heated to a temperature in the temperature range in which the alloy exists in xcex1+xcex2 phase and is then rapidly quenched.
In the above-described zirconium-base alloy according to the present invention, the region surrounded by the lines i)-v) on the rectangular coordinates defining the Fe content (X wt. %) of the zirconium-base alloy and the-mean size (Y nm) of precipitates in the zirconium-base-alloy is as shown in FIG. 1 or 2.
A nuclear reactor component according to the present invention comprises the above-described zirconium alloy.
The zirconium alloys of the present invention are excellent in both corrosion resistance and hydrogen absorption characteristics and are useful as materials for nuclear reactors. In particular, they are excellent as fuel cladding materials for use in nuclear fuel elements, and can increase the cladding life.